ML20027E155
| ML20027E155 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/03/1982 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| Shared Package | |
| ML20027B587 | List: |
| References | |
| IEB-80-04, IEB-80-4, TAC-46855, NUDOCS 8211120158 | |
| Download: ML20027E155 (11) | |
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y fiGVEMBER 3 1962 3NBolL Docket No. 50-312 7
Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S Street P. O. Box 15830 Sacramento, California 95813
Dear Mr. Mattimoe:
SUBJECT:
RANOIO SECO NUCLEAR GENERATING STATION - MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION We have completed our review of your responses to IE Bulletin 80-04, VAnalysis of a PWR Main Steam Line Break with Continued Feedwater Addition." Our Safety Evaluation Report (SER) and the Technical Evaluation Report (TER) by our consultants the Franklin Research Center is enclosed. From our review we have concluded thattfor the current Emergency Feedwater System, there is no potential 'fbr containment overpressurization and the reactivity increase analysis
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bounds the current design. However, the AFW pumps are not protected against runout. Theref. ore, please. demonstrate that the pumps will not be damaged by runout or describe the interim measures that'will be taken to prevent runout.
Your proposed upgraded Emergency Feedwater System, the EFIC System, directs feedwater to both steam generators and the flow of feedwater to both steam generators is not bounded by the asstsoptions in your FSAR analysis. Therefore, please provide an analysis of a Main Steam Line Break (HSLB) which determines the containment pressure response and reactivity increase for the proposed EFIC system. In addition in the EFIC system design, the AFW pumps are not protected from runout during a double steam generator blowdown. Therefore, please describe how you propose to prevent runout or protect the pumps against runcut for this case.
Please provide the above requested information within 45 days of receipt of this letter.
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OFFiClAL RECORD COPY usaro: mi-a=*o NRC FORM 318 00-80) NRCM Cao
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Mr. J. J. Mattimoe.
The reporting requirements contained in this letter affect fewer than ten respondents; therefore, ONB clearance is not required under P. L.96-511.
Sincerely,
'uuAGIRAL SIssgo gy YSEP. SIOLz'> '
John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
Enclosures:
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l NRC FORM 318 00-60) NRCM wo OFFiClAL RECORD COPY usom; issi..asseeo
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Sacramento Municipal Utility Rancho Seco, Docket No. 50-312 District ccw/ enclosure (s):
David S. Kaplan, Secretary and Christopher Ellison, Esq.
General Counsel Dian Grueuich, Esq.
Sacramento Municipal Utility California Energy Commission District 1111 Howe Avenue 6201 S Streat Sacramento, California 95825 P. O. Box 15830 Sacramento, California 95813 '
Ms. Eleanor Schwartz California State' OffJce Sacramento County 600 Pennsylvan'ia Avenue, S.E., Rm.' 201 Board of Supervisors Washington, D. C.
20003 827 7th Street, Room 424 Sacramento, California 95814 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Resident Inspector / Rancho Seco c/o U. S. N. R. C.
14410 Twin Cities Road Herald, CA 95638 1
Dr. Richard F. Cole Atomic Safety & Licensing Board Panel U.S. Nuclear Regulatory Commission
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Washington, D. C.
20555 1
Regional Radiation f.epresentative EPA Recion IX Mr. Frederick J. Shon 215 Fremont Street Atomic Safety and Licensing Board San Francisco. California 94111 Panel U.S. Nuclear Regulatory Commission Mr. Robert B. Borsum Washington, D. C.
20555 Babcock & Wilcox Nuclear Power Generation Division Elizabeth S. Bowers, Esq.
Suite 220, 7910 Woodmont Avenue Chairman, Atomic Safety and Bethesda, Maryland 20814 Licensing Board Panel U.S. Nuclear Regulatory Commission Thomas Baxter, Esq.
Washington, D. C.
20555 Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036 Herbert H. Brown, Esq.
Lawrence Coe Lanpher, Esq.
e Hill, Christopher and Phillips, P.C.
1900 M Street, N.N.
Atomic Safety' and Licensing Board Washington, D. C.
20036 Panel U.S. Nuclear Regulatory Commission Helen Hubbard Washington, D. C.
20555 P. O. Box 63 Sunol, California 94586
Sacramento Municipal Utility.
District Atomic Safety and Licensing Appeal Mr. Robert H. Engelken, Regional Administrator Board Panel U. S. Nuclear Regulatory Commission, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Washington, D. C.
20555 Walnut Creek, California 94596 Alan S. Rosenthal, Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Christine ii. Kohl Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20553 California Department of Health ATTN: Chief. Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 i
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SAFETY EVALUATION REPORT CONTAINMENT SYSTEMS BRANCH MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION RANCHO SECO NUCLEAR PLANT UNIT 1
.. e Docket No.: 50-312
1.0 INTRODUCTION
In the summer of 1979, a pressurized water reactor (PWR) Licensee submitted a report to the NRC that identified a deficiency in its original analysis of containment pressurization resulting from a postulated main steam Line break (MSLB).
A reanalysis of the containment pressure response fotLowing a MSLB was performed, and 1
it was determined that, if the auxiliary feedwater (AFW) system continued to supply feedwate,r at runout conditions to the st'eam generato r that had expe rienced the steam Line break, the containment design pressure would be exceeded in approximately 10 minutes.
In ot he r wo rds, the long-term blowdown of the water supplied by the AFW system had not been considered in the earli er analysis.
On Octobe r 1, 1979, the f o regoi ng inf ormation was provided to all holde rs of ope rating Licenses and construction permits in IE Information Notice 79-24 C23.
Another Licensee performed an accident analysis review pursuant to the inf ormation f urnished in the above cited notice and di scovered that, with offsite electrical power availabLe, the condensate pumps would f eed t he af fect ed steam generator at an excessive rate.
This excetsive feed had not been considered in the analysis of the postulated MSLB accident.
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A third Licensee informed the NRC of an error in the MSLB analysis for their plant.
For a zero or low power condition at the end of core life, the Licensee identified an incorrect postulation that the-startup feedwater control valves would remain positioned "as is" during the transient.
In reality, the startup feedwater control valves wilL ramp to 80% f.nl L od'en due to an override signal resulting from the Low steam' generator pressure reactor trip signal.
Reanalysis of the events showed that the rate of feedwater addition to the affected steam gene-rator associated with the opening of the startup valve would cause a rapid reactor cooldown and resultant reacto r-return-to powe r response, a condition which is beyond the plant's design basis.
FolLowing the identification of these deficiencies in the original MSLB accident analysi, the NRC issued IE ButLetin 80-04 on February 8, 1980.
This but Letin requi red alL licensees of PWRs and near-term PWR operating License applicants to do the fotLowing:
1.
Review the containment pressure response analysis to determine if the potential for containment ove rp r e s sur e in the event of a MSLB inside containment included the impact of runout flow from the auxiliary.feedwater system and the impact of other energy sources such as continuation of feedwater or condensate flow.
In your review, consider the ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
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Review your analysis of the reactivity increase which results from a MSLD inside or outside containment.
T.his review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position.
If your previous analysis'did not consider aLL'fpote$tial water sources (such as those Listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should include:
a.
The boundary conditions for the analysis, e.g.,
the end of Life shutdown margin, the moderator temperature l
coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc; b.
The most restrictive single active failure in the safety injection system and the effect of thdt failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system; I
l c.
The effect of extended water supply to the affected steam generator on the core criticality and return to power; and i
d.
The hot channel factors corresponding to the most reactive rod in the f',LLy withdrawn positions at the end of Life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient..
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If the potential f or containment overpressuri:ation exists or the reactor return-to power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action.
If the unit is operating, provide a descript. ion of any i n t.e ri m action that wi l, L' be,t ak e n un t il.
...a the proposed corrective action i s c omple t ed."
FolLowing the Licensee's initial response to IE ButLetin 80-04, a request for additional inf ormation was developed to obtain at L the information necessary to evaluate the Licensee's analysis.
The result s of our evaluation f or Rancho Seco Nuclear Plant, Rancho Seco (Rancho Seco) are provided below.
2.0 Evaluation Our consultant, the Franklin Research Center (FRC), has reviewed the submittats made by the Licensee in response to IE Butletin 80-04, and prepared the attached Technical Evaluation Report.
We t
have reviewed this evaluation and concur in its bases and findings.
l 3.0 conclusion i
Based on our review of the enclosed Technical Evaluation Report, the fotLoving conclusions are made regarding the postulated MSLB with continued feedwater addi tion f o r Rancho Seco:
1.
There is no potential for containment ov e rp r e s s ur i z a t io n r
l resulting from a MSLB with continued feedwater addition under the current AFW, system design because the main feed-l water system and auxiliary feedwater systems are isolated.
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2.
An analysis of the containment pressure response to a MSLB should be conducted for the EFIC system.
3.
The FSAR reactivity increase analysis bounds the current plant design.
4.
An analysis of the reactivity response to a MSLB, should be con-ducted for tNe E FIC systed/
l-5.
Under the current plant design, the AFW pumps are not protected against runout flow. Therefore the licensee should demonstrate that the AFW pumps will not be damaged during runout flow conditions or describe the interim actions that will be taken to protect the AFW pumps against runout flow.
6.
Under the design of the EFIC system, the AW pumps are protected against runout except during a double steam generator blowdown.
Therefore the licensee should protect the pumps against runout or prevent runout during a double steam generator blowdown.
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4.0 REFERENCES
1.
" Analysis of a PWR Main Steam Line Break wit Continued Feedwater Addition" NRC Office of Inspection and Enforcement, February 8,1980 IE ButLetin 80-04 2.
"overpressuri zation of the Containment of a PWR* Plant after a Main Stear Line Break".'<'
ff.
NRC Office of Inspection and Enforcement, Octobe r 1,1979 IE Information Notice 79-24 3.
J.J. Mattimoe (SMUD)
Letter to R. H. Enge lk en (NRC, Region V)
Subject:
IE ButLetin 80-04 May 6,1980 4.
J.J. Mattimoe (SMUD)
Letter to J.
F.
Stolz (NRR)
Subject:
Response to Request for Additional Information, l
IE Bu'lletin 80-04 May 28,1982 l
5.
Rancho Seco Nuclear Generating Statoin Final Safety Analysis Report, through Amendment 27 Sacramento Municipal Utility District i
6.
Technical Evaluation Re,por.t "PWR Main Steam Line Break with Continued F,eedwater Addition - Review of Acceptance Criteria Franklin Research Center November 17, 1981 7.
" Criteria f or Protection S ystems for Nuclear Power Generating Stations" Institute of Electrical and Electronics Engineers, New York, N Y, 19 71 l
Standard Review Plan, Section 4.2
" Fuel S ystem Design" l
NRC, J uly 1981 1
9.
Standard Review Plan, S e ct i on 15.1.5 l
" Steam S ystem Piping Failures Inside and Outside of Contain-l ment (PWR)"
NRC, J uly 1981 NUREG-0800.
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" Criteria for Accident Monitoring Functions in Light-Water Cooled Reactors" Ame ri can Nu clea r Societ y, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980 11.
" Instrumentation f or Light-Water-Cooled Nuclea r Power P Lants to Assess Plant and Environs Conditions During and FolLowing an Accident" i
Rev. 2 NRC, December 1980 Regulatory Guide 1.97 12.
" Single Failure Criteria for PWR Fluid S ystems" American Nucelar Society., Hinsdale, IL, June 1976 ANS-51.7/N658-1976 13.
"Qualit y G roup Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power P Lants" Rev. 3 NRC, February 1976 Regulatory Guide 1.26 14.
" Interim Staf f Position en Environmental Qualification of S a f et y-Relat ed E lect ri ca L Equipment" Rev. 1 NRC, J uly 1981 NUREG-0588 15.
W.
S. Bossenmaier (SMUD)
Letter to R.
W. Reid (NRR)'
Subject:
Upgraded Auxiliary F eedwat e r S yst'em D esi gn S ub-mittal November 17, 1980 16.
J.J. Mattimoe (SMUD)
Letter to J.
F.
Stotz (NRR)
Subject:
TMI Ac t ion P Lan I t ems II.E.1.2 a n d II.K.2.10 September 8, 1981.
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